1. FIELD OF THE INVENTION
This invention relates to nuclear fuel assemblies and to nuclear reactor cores containing such assemblies, and is particularly concerned with such fuel assemblies for a boiling water reactor (BWR).
2. DESCRIPTION OF THE PRIOR ART
FIG. 7 of the accompanying drawings shows in horizontal section an arrangement of four fuel assemblies in a current BWR core. Specifically, a fuel bundle, which is constructed by arranging fuel rods 2 in a square lattice or array, is in the middle of a unit cell 22 of the core, and a channel box 23, positioned outside the fuel bundle, defines a water gap region around the fuel assembly outside the channel box for flow of light water as a coolant without boiling. Due to the presence of a space to allow a cross-shaped control rod 24 to be inserted, this water gap region is composed of gaps 25, at which the control rod 24 is inserted (herein called a control rod region) and gaps 26 at which the control rod 24 is not inserted nor withdrawn (herein called a non-control rod region).
In order to reduce the power cost for a light-water type reactor, it is effective to reduce the fuel cycle cost. One of the methods is to increase the average discharge exposure. Generally, the more the discharge exposure is increased, the more the average enrichment of the fuel assembly is increased. As the fuel enrichment becomes higher, the reactivity in the cold state of the core (about 20.degree. C.) becomes higher, so that the difference between the reactivities at the hot state (which means that the coolant temperature is about 280.degree. C.) and at cold becomes larger (that is to say, the reactivity difference between hot and cold increases). With a high reactivity, the control rods are used to control so as to balance the amounts of production and absorption of neutrons. When inserted in the hot state and in the cold state, the control rod is present at one side of the fuel assembly as shown in FIG. 7.
The neutron absorber B.sub.4 C in the control rod decreases reactivity, especially by absorbing thermal neutrons. Even if the relative enrichment distribution of the fuel assembly is identical, when the enrichment is increased, the effect of absorption of the thermal neutrons by the control rod is decreased. Therefore, the so-called control rod worth at the cold is decreased.
The control rod worth is the value of reactivity worth of the control rod, which is proportional to the square of the neutron flux at the position of the control rod.
In the case of the core for a high-burn-up, therefore the cold shut-down margin has a tendency to become small (that is to say, undesirably the margin may easily become critical at the shut-down of the reactor).
In order to improve the reactor cold shut-down margin, the following methods can be considered:
(1) Enhancing the control rod worth to increase the difference between the reactivities at cold with and without the insertion of the control rod; and
(2) Decreasing the difference between the reactivities at the hot and at the cold states.
The BWRs being run at present have cores of two different types, i.e. the C-lattice core, in which the water gap regions 25 and 26 have an equal area, as shown in FIG. 7, and the D-lattice core in which the water gap 25 which receives the control rod 24 is wider than the water gap 26 which does not receive a control rod (non-control rod region).
FIG. 8 of the accompanying drawings exemplifies the method of enhancing control rod worth disclosed in JP-A-61-275696, in which the enrichment with fissile material of fuel rods 32 at the sides of the fuel rod array adjacent the cross-shaped control rod 31 is higher than the average enrichment per fuel rod of the fuel assembly. This arrangement can indeed achieve improvement of the control rod worth, because the control rod when inserted into the core is adjacent the fuel regions of highest enrichment. However, when this is applied to a D-lattice type core, during operation when the control rods are absent, the regions of high enrichment are next to wide water gaps providing a high neutron moderation and consequently the local power peaking factor at these regions of high enrichment is likely to be large. The local power peaking factor is defined as the ratio of maximum local power in the fuel assembly, as seen in cross section, relative to the average value of the power. In a reactor having a small local power peaking factor, power output is flattened as is desirable.
FIG. 9 of the accompanying drawings shows another method disclosed in JP-A-63-98590 for enhancing the control rod worth in the cold condition of the reactor. The fuel assembly is divided into four sub-assemblies 43,44, and the average enrichment of the sub-assembly 44 which is closest to the control rod 45 is higher than in the other three sub-assemblies 43. In this fuel assembly, a cross-shaped water channel is arranged at the center of the fuel assembly to divide it into the four sub-assemblies 43 and 44.
Another method of improvement of control rod worth is disclosed in JP-A-64-31091, in which the fuel assembly includes at least one water rod. The water rod or rods is shifted in the fuel rod array from the conventional central position towards the corner of the array at which the two sides adjacent the non-control rod region meet. This has the effect of shifting some of the fuel rods containing fissile material towards the control rod region, thus improving the control rod worth. This prior art document also discloses a conventional distribution of the fissile material among the fuel rods in the array, in which, when the array is notionally divided into four regions, and excluding from any of those regions rods lying on the diagonal line joining the opposite corners at which the control rod region meets the non-control rod region, these four regions of the array being
a: the row and column adjoining the control rod region; PA1 b: the fuel rods between the region a and the said diagonal line; PA1 c: the fuel rods lying between the region d and said diagonal line, PA1 d: the row and column adjoining the non-control rod regions; PA1 a: the rods in the row and column of the array adjoining said first sides; PA1 b: the rods lying between said region a and said diagonal line; PA1 c: the rods lying between said region d and said diagonal line, PA1 d: the rods in the row and column of the array adjoining said second sides and;
the average concentration of fissile material per fuel rod in the region c is lower than in the region b. A fuel distribution of this kind is given in FIGS. 10 and 14 of the drawings of the present application. FIG. 10 show, in horizontal section a 8.times.8 square array of fuel rods in a fuel assembly and FIG. 14 a 9.times.9 square array of fuel rods in a fuel assembly. The symbols A,B,C etc. used in FIG. 10 are explained below in connection with FIG. 1 and Table 1, in which they have the same meaning. Likewise the symbols A',B',C'. . . G' have the same meaning as in FIG. 3 and Table 2 below, and the fuel enrichment of rods H' of FIG. 14 is 4.30 wt. % (no Gd content). Regions a,b,c,d defined above are indicated by reference numerals 4,5,6,7 respectively in FIG. 10. The exact fuel distributions of FIGS. 10 and 14 are not disclosed in the prior art, and are included here for the purposes of comparison with the fuel rod assemblies embodying the invention discussed below. These fuel distributions are called representative cases, below.
The reason for this conventional fuel distribution in a core of the D-lattice type is to achieve a flattened power distribution in the array, i.e. a low local power peaking factor. The regions a and d are close to water gaps, where the moderating effect of the water results in a higher neutron flux, leading to higher output in these regions. Therefore the fuel concentration is lower in these regions. In the D-lattice core, the control rod gap adjoining the region a is wider than the non-control gap adjoining the region d. Therefore the region b is closer to a wider water gap than the region c. To achieve a flattened power output, the region b is given a lower average fuel concentration per rod than the region c, to compensate for the moderating effect of the water in the wider water gap.
In the fuel assemblies of JP-A-64-31091 mentioned above, the effect of the shifting of the water rods into the region c is to disturb substantially the flattened power distribution achieved with the conventional fuel distribution just described. The moderating effect of the water rod or rods in the region c makes the region c liable to have a high power output, leading to a high local power peaking factor.
EP-A-284016 and corresponding US-A-4876062 are primarily concerned with the shape of water rods in nuclear fuel assemblies, but disclose in FIG. 22 a fuel assembly having a water rod which is located partly in the region c mentioned above, and having a fuel distribution in which the average concentration per rod in the region b is greater than in the region c by about 3.5%. In particular, fuel rods containing burnable poison are located close to the water rod. In connection with the location of the water rod and also the specific cross-sectional shape of the water rod of that FIG. 22, it is mentioned that the distribution of neutron fluxes can effectively be made even. No reason is given for this particular pattern of fuel enrichment.